ML082820047

Issuance of Amendments Regarding Revision 1 to DPC-NE-1005-P, Nuclear Design Methodology Using CASMO-4/Simulate -3 MOX

ML082820047
Person / Time
Site:	Catawba
Issue date:	11/12/2008
From:	Stang J Plant Licensing Branch II
To:	Morris J Duke Energy Carolinas
Stang J, NRR/DORL 415-1345
References
TAC MD7407, TAC MD7408
Download: ML082820047 (17)

Text

November 12, 2008 Mr. J. R. Morris Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REVISION 1 TO DPC-NE-1005-P, NUCLEAR DESIGN METHODOLOGY USING CASMO-4/SlMULATE-3 MOX (TAC NOS.

MD7407 AND MD7408)

Dear Mr. Morris:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 246 to Renewed Facility Operating License NPF-35 and Amendment No. 239 to Renewed Facility Operating License NPF-52 for the Catawba Nuclear Station, Units 1 and 2, respectively. The amendments authorize changes to the licensing bases and final updated safety analysis report for the Catawba Nuclear Power Station, Units 1 and 2, concerning Revision 1 to DPC-NE-1005-P, Nuclear Design Methodology Using CASMO-4/SlMULATE-3 MOX, in response to your application dated November 12, 2007, as supplemented by letter dated April 8, 2008.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/ LOlshan for John Stang, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413 and 50-414

Enclosures:

1. Amendment No. 246 to NPF-35

2. Amendment No. 239 to NPF-52

3. Safety Evaluation cc w/encls: Distribution via ListServ

Mr. J. R. Morris Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REVISION 1 TO DPC-NE-1005-P, NUCLEAR DESIGN METHODOLOGY USING CASMO-4/SlMULATE-3 MOX (TAC NOS.

MD7407 AND MD7408)

Dear Mr. Morris:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 246 to Renewed Facility Operating License NPF-35 and Amendment No. 239 to Renewed Facility Operating License NPF-52 for the Catawba Nuclear Station, Units 1 and 2, respectively. The amendments authorize changes to the licensing bases and final updated safety analysis report for the Catawba Nuclear Power Station, Units 1 and 2, concerning Revision 1 to DPC-NE-1005-P, Nuclear Design Methodology Using CASMO-4/SlMULATE-3 MOX, in response to your application dated November 12, 2007, as supplemented by letter dated April 8, 2008.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/ LOlshan for John Stang, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413 and 50-414

Enclosures:

1. Amendment No. 246 to NPF-35

2. Amendment No. 239 to NPF-52

3. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION: Public RidsAcrsAcnw_MailCTR Resource LPL2-1 R/F GHill, OIS(4 hard copies)

RidsNrrDorlLpl2-1 Resource RidsNrrDssSnpb (AMendiola)

RidsNrrPMJStang (hard copy) RidsRgn2MailCenter (JMoorman)

RidsNrrLAMO=Brien (hard copy) RidsNrrDorlDpr RidsOgcRp Resource RidsNrrDssSnpbAAttard, NRR/DSS/SNPB RidsNrrSnpbJWoods Amendment No.: ML082820047 *SE transmitted by memo dated 7/15/08.

OFFICE NRR/LPL2-1/PM NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/SNPB/BC OGC NRR/LPL2-1/BC NAME JThompson JStang MO=Brien AMendiola AJones MWong DATE 10/27/08 10/27/08 10/27/08 7/15/08* 11/10/08 11/ /08 OFFICIAL RECORD COPY

DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 246 Renewed License No. NPF-35

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-35 filed by the Duke Energy Carolinas, LLC, acting for itself, and North Carolina Electric Membership Corporation (licensees), dated November 12, 2007, as supplemented by letter dated April 8, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is hereby amended as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 246, which are attached hereto, are hereby incorporated into this license. Duke Energy Carolinas, LLC, shall operate the facility in accordance with the Technical Specifications.

3. Further, Renewed Facility Operating License No. NPF-35 is hereby amended to authorize a change to the Updated Final Safety Analysis Report (UFSAR) to modify sections of the UFSAR as set forth in the license amendment application dated November 12, 2007, supplemented by letter dated April 8, 2008, and evaluated in the safety evaluation dated November 12, 2008. The wording of UFSAR Section 4.3.3, Analytical Methods, will be modified to indicate that the CASMO-4/SIMULATE-3 MOX nuclear design methodology will be used to model cores with fuel containing gadolinia.

The wording of UFSAR Section 4.3.6, References, will be modified to replace references to DPC-NE-1005-P-A, Revision 0, with Revision 1 of this topical report.

Other changes to the UFSARs include minor editorial changes and typographical error corrections. The licensee shall update the UFSAR by adding a description of the change, as authorized by this amendment, and in accordance with 10 CFR 50.71(e).

4. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Melanie C. Wong, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-35 Date of Issuance: November 12, 2008

DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 239 Renewed License No. NPF-52

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Renewed Facility Operating License No. NPF-52 filed by the Duke Energy Carolinas, LLC, acting for itself, North Carolina Municipal Power Agency No. 1 and Piedmont Municipal Power Agency (licensees), dated November 12, 2007, as supplemented by letter dated April 8, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (I) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is hereby amended as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 239, which are attached hereto, are hereby incorporated into this license. Duke Energy Carolinas, LLC, shall operate the facility in accordance with the Technical Specifications.

3. Further, Renewed Facility Operating License No. NPF-52 is hereby amended to authorize a change to the Updated Final Safety Analysis Report (UFSAR) to modify sections of the UFSAR as set forth in the license amendment application dated November 12, 2007, supplemented by letter dated April 8, 2008, and evaluated in the safety evaluation dated November 12, 2008. The wording of UFSAR Section 4.3.3, Analytical Methods, will be modified to indicate that the CASMO-4/SIMULATE-3 MOX nuclear design methodology will be used to model cores with fuel containing gadolinia.

The wording of UFSAR Section 4.3.6, References, will be modified to replace references to DPC-NE-1005-P-A, Revision 0, with Revision 1 of this topical report.

Other changes to the UFSARs include minor editorial changes and typographical error corrections. The licensee shall update the UFSAR by adding a description of the change, as authorized by this amendment, and in accordance with 10 CFR 50.71(e).

4. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Melanie C. Wong, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-52 Date of Issuance: November 12, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 246 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 239 RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License Page License Page License No. NPF-35, page 4 License No. NPF-35, page 4 License No. NPF-52, page 4 License No. NPF-52, page 4

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 246 TO RENEWED FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO. 239 TO RENEWED FACILITY OPERATING LICENSE NPF-52 DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

By application dated November 12, 2007 (Reference 1), as supplemented by letter dated April 8, 2008 (Reference 2), Duke Energy Carolinas, LLC (Duke, the licensee), requested changes to the Renewed Facility Operating Licenses (FOLs) and Updated Final Safety Analysis Reports (UFSARs) for the Catawba Nuclear Station, Units 1 and 2 (Catawba). The supplement dated April 8, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff=s original proposed no significant hazards consideration determination as published the Federal Register on January 29, 2008 (73 FR 5218).

The proposed changes address the intentions of Duke to use the gadolinia integral fuel burnable absorber in future reload cores at the McGuire Nuclear Station (McGuire) and Catawba units. To develop the methodology that the licensee will use to perform nuclear design calculations for reactor cores containing gadolinia fuel, Duke has submitted Revision 1 of the topical report (TR) DPC-NE-1005-P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX (Reference 3), for U.S. Nuclear Regulatory Commission (NRC) staff review.

Duke previously submitted Revision 0 of report DPC-NE-1005-P (Reference 4), which describes the CASMO-4/SIMULATE-3 MOX methodology used to develop uncertainty factors for reload design calculations. However, the application of this methodology was limited to the analysis of reactor cores containing low enriched uranium (LEU) fuel, and for use of up to four mixed oxide (MOX) lead test assemblies in one of the Catawba units. The presence of gadolinia bearing fuel rods was not considered. The introduction of gadolinia fuel is considered a significant fuel design change, and thus requires further review and validation of the methodology by NRC staff.

The licensees transition to core designs containing gadolinia burnable absorbers will occur simultaneously with the transition from the Westinghouse Electric Company Robust Fuel

Assembly (RFA) to the Advanced Mark-BW (ABW) fuel design of AREVA NP, Inc. (AREVA NP).

The complete fuel transition will require additional submittals from the licensee outlining the fuel transition methodology and requests for changes to Catawba and McGuire technical specifications. The future submittals will detail the methodologies for performing core reload design, fuel assembly mechanical and thermal hydraulic analyses and UFSAR Chapter 15, non-loss-of-coolant accident transient and accident analyses, related to the transition to AREVA NP ABW fuel. However, this Safety Evaluation (SE) only addresses the application of the licensees nuclear design methodology to the analysis of reactor cores containing gadolinia fuel.

2.0 REGULATORY EVALUATION

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 50.90, the licensees license action request (LAR) submittal must fully describe the changes desired.

The provisions of 10 CFR 50.59(c)(2)(viii) require that A licensee shall obtain a license amendment pursuant to Sec. 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses. Duke has submitted proposed changes to the FOLs and UFSARs. These changes are intended to extend the nuclear design methodology to incorporate the methods described in Revision 1 of TR DPC-NE-1005-P. Dukes submittal includes a description of the proposed changes, as well as marked-up pages from the existing FOLs and UFSARs. As indicated above, the proposed change requires prior approval by the NRC.

Assessments of the proposed changes are provided as follows:

• The proposed changes to the Catawba and McGuire UFSARs consist of modifications to the wording in Section 4.3.3, Analytical Methods, and Section 4.3.6, References. The changes indicate that the CASMO-4/SIMULATE-3 MOX nuclear design methodology is used to model cores with fuel containing gadolinia. References to DPC-NE-1005-P-A, Revision 0 in the UFSARs are replaced with references to Revision 1 of the TR. Other changes to the UFSARs include minor editorial changes and typographical error corrections.

• The inclusion of Revision 1 to DPC-NE-1005-P-A in the references to the UFSARs requires additional review and assessment of the contents of this report. The technical evaluation of Revision 1 to DPC-NE-1005-P-A is contained in Section 3.0 of this SE.

• The nuclear design review of fuel assemblies, control systems, and reactor core is carried out to aid in confirming that fuel design limits will not be exceeded during normal operation or anticipated operational transients. The NRC staff acceptance criteria are based on Chapter 4.3, Nuclear Design, of NUREG-0800, Revision 2, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants.

3.0 TECHNICAL EVALUATION

The proposed UFSAR changes in this submittal involve extending the CASMO-4/SIMULATE-3 MOX methodology to be used for analysis of reactor cores containing gadolinia. The marked-up UFSAR pages refer to Revision 1 of DPC-NE-1005-P-A for the technical description of the

methodology. Appendix B of Revision 1 to this report describes the methodology that is used to determine the uncertainty of CASMO-4/SIMULATE-3 MOX predictions for cores containing LEU fuel with gadolinia fuel. Other revisions to the TR include changes to the original content to address the inclusion of the gadolinia methodology, corrections of typographical errors, and editorial changes to add clarity. The technical evaluation in this SE focuses on the contents of Appendix B to DPC-NE-1005-P-A.

The CASMO-4/SIMULATE-3 MOX code system, as it pertains to this evaluation, is applied to reactor cores containing LEU fuel with gadolinia. Applications with MOX fuel are not considered. Accordingly, the code system is hereinafter referred to as CASMO-4/SIMULATE-3.

Appendix B to DPC-NE-1005-P-A presents benchmark comparisons between CASMO-4/SIMULATE-3 calculations and data from both operating reactor cores and critical experiments. These benchmark comparisons are the basis for determining the accuracy of CASMO-4/SIMULATE-3 predictions for critical boron concentrations, isothermal temperature coefficients, control rod bank worth, and assembly and pin power distributions for cores containing gadolinia fuel. The methodology and results of the benchmarks from Appendix B are evaluated in Sections 3.1 through 3.5 below.

3.1 Computer Codes and Models The codes that the licensee used for benchmarking comparisons with reactor cores containing gadolinia are: CASMO-4, SIMULATE-3, and CMS-LINK. The CASMO-4 code is a multi-group, two-dimensional transport theory model with a microscopic depletion model for burnable absorbers. CASMO-4 is used to calculate lattice physics parameters, including cross sections, pin power distributions and other nuclear data, which are used as inputs to SIMULATE-3. The SIMULATE-3 code is a two-group, three-dimensional coarse-mesh nodal diffusion theory simulator. SIMULATE-3 combines the nodal solution with the heterogeneous lattice solution from CASMO-4 to calculate pin power distributions. To model reactor cores containing gadolinia, no modifications to the nodal solution or pin power reconstruction routines were necessary. The other code that is used with this package is CMS-LINK, which processes data from CASMO-4 to produce multi-dimensional tables for input to SIMULATE-3.

3.2 Sequoyah Benchmark Analyses The licensee used the CASMO-4/SIMULATE-3 code system to calculate predicted reactivity parameters and fuel assembly power distributions for comparisons to measured data from the Sequoyah Unit 2 reactor cores containing gadolinia fuel. Measured data for critical boron concentration, isothermal temperature coefficient, and control rod bank worth was compiled from startup physics testing and routine monitoring during the operation of Sequoyah Unit 2 Cycles 9 through 13. Also, core power distribution measurements for these cycles were taken at regular intervals with the Sequoyah incore system. The licensee compared these measured data to the predicted values in order to assess the uncertainty of the CASMO-4/SIMULATE-3 models of reactor cores containing gadolinia fuel.

Cycles 9 through 13 of the Sequoyah Nuclear Plant (Sequoyah), Unit 2, were selected for benchmark comparisons because of the similarities between Sequoyah Unit 2 and the McGuire and Catawba reactors. The licensee considers the loading patterns and fuel assembly designs

used in these cycles to be applicable to future McGuire and Catawba core designs containing gadolinia. The Sequoyah benchmark cycles included both transition and full core designs containing AREVA NPs Mark-BW fuel with gadolinia.

3.2.1 Critical Boron Concentration Comparisons Critical boron concentration measurements for Sequoyah Unit 2 Cycles 9 through 13 were taken during start-up physics testing at the beginning of each cycle and throughout full-power operation by sampling the reactor coolant. The measurements made during start-up physics testing were taken at beginning of cycle (BOC), hot zero power (HZP) conditions, with all rods out of the core (ARO), peak samarium, and no xenon. The full-power measurements were taken at or near hot full-power (HFP) nominal conditions at several burnups throughout the cycle. Full-power measurements were corrected for B-10 depletion and from the measured rodded condition to the ARO condition.

Using the CASMO-4/SIMULATE-3 models, the licensee calculated a predicted value of critical boron concentration for each measurement from Sequoyah Unit 2, Cycles 9 through 13. The predicted critical boron concentrations agree well with the measurements. The NRC staff has previously reviewed critical boron concentration comparisons from CASMO-4/SIMULATE-3 MOX predictions to McGuire and Catawba measurements for reactor cores not containing gadolinia (Reference 5). The results of these previous comparisons are similar to the results of the comparisons with Sequoyah cores containing gadolinia fuel.

3.2.2 Isothermal Temperature Coefficient Comparisons The isothermal temperature coefficient (ITC) was measured at BOC, HZP, ARO conditions during startup physics testing for Sequoyah Unit 2, Cycles 9 through 13. The licensees predictions of ITC using CASMO-4/SIMULATE-3 deviate from the measured values by an acceptable amount. The agreement between predicted and measured ITC values for the Sequoyah cycles is slightly better than the agreement of previous predictions to McGuire and Catawba non-gadolinia cores.

3.2.3 Control Rod Worth Comparisons During Sequoyah Unit 2 startup testing for Cycles 9 through 13, control rod bank worth measurements were made at BOC, HZP, peak samarium and no xenon conditions. The Sequoyah control bank worth measurements were performed using the Rod Swap technique, as described in Appendix B to DPC-NE-1005-P. Comparisons to CASMO-4/SIMULATE-3 predictions were made for each of the individual bank worths and for the sum of bank worths for each cycle. All predicted values are in good agreement with the measured control rod bank worths. The results of these comparisons are also similar to previous control rod bank worth comparisons for McGuire and Catawba cores not containing gadolinia.

3.2.4 Fuel Assembly Power Distribution Analysis and Uncertainty Factors Core power distributions were measured at regular intervals during operation using the Sequoyah moveable incore fission chamber system. During Cycles 9 through 13, 88 power distribution measurements were made to determine the measured assembly peaking factors FH, Fq, and Fz. The licensee used SIMULATE-3 to model the reactor conditions for all 88 power distribution measurements and produce predicted values for the assembly peaking

factors. The uncertainties between the predicted and measured values are characterized by assembly uncertainty factors.

Duke defines assembly uncertainty factors, referred to as Observed Nuclear Reliability factors (ONRFs), according to the following expression:

ONRF = 1 - bias + Kaa where the bias is the mean of predicted minus measured values, and Kaa is the statistical deviation of the bias. (The a subscript is used to represent assembly-averaged values.) The a is the standard deviation of the bias distribution, and the Ka factor is determined from a 95%

one-sided upper tolerance interval with a 95% confidence level, as described in References 6 and 7. This determination of Kaa requires that the data pass a test for normality (Reference 8) at the 1% level of significance. If the data fail the normality test, then a conservatively large value is assigned to Kaa using a nonparametric evaluation from References 6 and 9. Note that all data used in these calculations were deemed to be normally distributed. The statistical methods used to calculate the ONRFs have been reviewed and are found to be acceptable.

The calculated ONRF values for assembly FH, Fq, and Fz indicate good agreement between measured power distributions and those predicted with SIMULATE-3. The ONRFs from this comparison to Sequoyah data have values similar to the ONRF values that the licensee calculated for McGuire and Catawba cores with only LEU fuel, not containing gadolinia.

The licensee intends to use the ONRFs as the basis for developing power peaking uncertainties for proposed gadolinia core designs at the McGuire and Catawba units. The complete power distribution uncertainties incorporate the bias and uncertainty of the assembly average power distributions along with the uncertainty from predictions of fuel pin power distributions. The licensees determination of uncertainty in the pin power distribution is discussed in Section 3.3 below.

3.3 Pin Power Uncertainty Factor To determine the accuracy of the CASMO-4 and SIMULATE-3 models for predicting pin power distributions in reactor cores containing gadolinia, the licensee compared these models to results from the Babcock & Wilcox (B&W) Urania Gadolinia critical experiments (Reference 10).

These benchmark comparisons were used to develop uncertainty factors for both LEU fuel pin powers and gadolinia fuel pin powers. The LEU pin power uncertainty was determined by direct comparison of SIMULATE-3 predictions with measurements from the critical experiments. For the uncertainty in gadolinia fuel pin power, the licensee employed an alternate approach. The B&W critical experiment measurements were taken at, or near, BOC conditions, where the gadolinia pin power density is non-limiting and significantly lower than that for the LEU pins.

The licensee based the gadolinia pin power uncertainty on a combination of comparisons to B&W critical experiment data and an evaluation of a series of theoretical infinite lattice 2x2 colorset calculations at different burnups.

3.3.1 LEU Pin Power Uncertainty The licensee determined the LEU pin power uncertainty by modeling the power distributions

from the B&W critical experiments for core configurations 5, 14, and 20, containing gadolinia fuel. LEU pin power distributions were calculated separately using CASMO-4 and SIMULATE-3, and uncertainties were developed for each code. The uncertainty is based on the predicted minus measured percent error and on the Kaa value, which is described in Section 3.2.4 of this SE and derived from References 6 and 7. The CASMO-4 and SIMULATE-3 predictions produce similar pin power uncertainty values, with the SIMULATE-3 uncertainty being slightly larger than the CASMO-4 uncertainty. Both uncertainties agree very well with the CASMO-4 and SIMULATE-3 LEU pin power uncertainties from Revision 0 of the TR (Reference 4), which compared predictions to B&W critical experiments with non-gadolinia cores.

The licensee uses the SIMULATE-3 LEU pin power uncertainty value in the final calculation of combined power distribution uncertainty factors (discussed in Section 3.4 below). The licensee does not provide a justification for using the SIMULATE-3 uncertainty rather than the CASMO-4 uncertainty, but this decision is acceptable since the SIMULATE-3 uncertainty is larger than the CASMO-4 uncertainty.

3.3.2 Gadolinia Pin Power Uncertainty The first component that the licensee used in the determination of the gadolinia pin power uncertainty was the benchmark comparison of B&W critical experiment data to predicted values.

The licensee used the CASMO-4 code to calculate the pin power distribution for the gadolinia fuel rods from B&W critical experiment core configurations 5, 14, and 20. The predicted powers were compared to measured data to find the bias (i.e. the mean difference of predicted minus measured values) and the standard deviation.

The bias and standard deviation were divided by the average gadolinia pin power to find the percent uncertainty. The average gadolinia pin power, as measured in the B&W critical experiments, was quite low. Using such a low value in these calculations is not meaningful.

The low gadolinia pin power, measured at BOC conditions in the critical experiments, is not representative of the higher pin powers that are reached after the gadolinia is depleted. The licensee chose a conservative value to use for the average gadolinia pin power in calculating the percent uncertainty. The value that the licensee chose is acceptable and recognized as conservatism in the uncertainty calculation.

The gadolinia pin power data from the B&W critical experiments were demonstrated to be normally distributed by the W test (Reference 8). A K-factor for a 95/95 upper tolerance was applied to the statistical uncertainty (as described in Reference 7). The NRC staff reviewed the data sets utilized for all the comparisons, as well as the methodology used in the uncertainty analysis, and determined that the results for the CASMO-4 gadolinia pin power uncertainty are acceptable (Reference 2).

The licensee determined that a comparison of the CASMO-4 predicted values to the B&W critical experiment data was, by itself, not sufficient to establish a gadolinia pin power uncertainty. The B&W critical experiment data taken at BOC conditions are non-limiting for gadolinia fuel pins. The gadolinia pin power is of most concern after the gadolinia is depleted and the gadolinia pin power approaches, or exceeds, the assembly average power. To resolve the uncertainty for this burnup range, the license performed a series of theoretical infinite lattice 2x2 colorset calculations with CASMO-4 and SIMULATE-3. The results of the two codes were compared to characterize the SIMULATE-3 to CASMO-4 pin power reconstruction uncertainty.

The licensee selected a diverse set of 2x2 fuel assembly loading patterns for the infinite lattice calculations with SIMULATE-3 and CASMO-4. The licensee evaluated 11 cases with different combinations of burned fuel and feed fuel assemblies with varying numbers of gadolinia fuel pins with concentrations up to 8.0 weight percent Gd2O3. Each colorset was modeled for a number of different burnups up to approximately 20 GWD/MTU, and gadolinia depletion was considered.

The licensee based the SIMULATE-3 gadolinia pin power reconstruction uncertainty on comparisons between the CASMO-4 and SIMULATE-3 power distributions for the 2x2 colorset cases described above. Both code calculations were normalized to an average assembly power of 1.0; therefore, the mean difference between the two predictions was 0. So, the gadolinia pin power uncertainty was based on the broadness of the distribution. The data set was tested for normality with the D test (Reference 8) and found to be not normal.

Consequently, the uncertainty was based on a non-parametric evaluation of the data set.

References 6 and 9 were consulted to determine the 95/95 one sided tolerance for the data.

This value was taken to represent the SIMULATE-3 to CASMO-4 gadolinia pin power reconstruction uncertainty.

The licensee calculated the total gadolinia pin power uncertainty by combining the SIMULATE-3 to CASMO-4 uncertainty with the K value from the B&W critical experiment comparison with CASMO-4. These values were combined by taking the square-root of the sum of the squares.

The bias term from the CASMO-4 comparison was also added to the uncertainty. The combination of these terms yields the total gadolinia pin power reconstruction uncertainty for SIMULATE-3. This methodology is deemed acceptable for producing a suitable value of gadolinia pin power uncertainty.

3.4 Statistically Combined Power Distribution Uncertainty Factors The licensee has defined power distribution uncertainty factors to be applied to peaking factors for design of reload cores and for surveillance of operating cycles. These uncertainty factors, referred to as statistically combined uncertainty factors (SCUFs), combine the inter-assembly power uncertainty and the intra-assembly pin power uncertainty. The SCUF is given by SCUF = 1 - bias + ( K a a ) 2 + ( K p p ) 2 where Kaa is the statistical deviation of the measured-to-predicted comparisons from the Sequoyah assembly average power distribution measurements. The Kpp term is the statistical deviation of the pin power distribution comparisons. The Kpp term combines the SIMULATE-3 to CASMO-4 uncertainty and the uncertainty from B&W critical experiment comparison to CASMO-4 predictions. The bias is given by the sum of the bias terms from the Sequoyah assembly power benchmark and the B&W critical experiment pin power benchmark.

The SCUF is calculated for each of the power distribution peaking factors FH, Fq, and Fz. The SCUFs are determined for LEU fuel and gadolinia fuel separately. These factors are applied to core reload designs and to surveillance tests to assure a conservative evaluation of fuel pin performance. The values of the uncertainty factors are reasonable, and the licensees methodology to determine the SCUF values is acceptable.

3.5 Summary of Assessment of Revision 1 to DCP-NE-1005-P-A The licensee intends to use the CASMO-4/SIMULATE-3 code system for reload design analyses for reactor cores containing gadolinia for McGuire and Catawba. To qualify this code system, the licensee has performed a series of benchmark comparisons. Reactivity and assembly power distribution predictions were compared to data from five Sequoyah Unit 2 fuel cycles. Fuel pin power distributions were compared to measurements from the Babcock &

Wilcox Urania Gadolinia critical experiments (Reference 10). The comparisons demonstrate the capability of the CASMO-4/SIMULATE-3 code system to adequately reproduce reactivity and power distribution calculations for reactor cores containing gadolinia.

The CASMO-4-based SIMULATE-3 predictions of reactivity parameters were compared to measurements from Sequoyah Unit 2 with acceptable accuracy. Comparisons were made for critical boron concentrations, both at BOC, HZP, and HFP conditions. ITC and control rod worth comparisons were made for BOC HZP conditions. All deviations between measurement and prediction produced similar results to those in the licensees prior submittal (Reference 5) for non-gadolinia cores.

CASMO-4 based SIMULATE-3 calculations were also used to predict the assembly average power distributions from the Sequoyah Unit 2 cycles containing gadolinia fuel. The calculated uncertainty factors for assembly FH, Fq, and Fz indicate good agreement between measured power distributions and the SIMULATE-3 predictions.

The pin power distribution uncertainty for LEU fuel rods was resolved by comparing CASMO-4 and SIMULATE-3 predictions to data from the B&W Gadolinia critical experiments. For the gadolinia fuel pins, the determination of pin power uncertainty was based on two inputs: 1) the comparison of CASMO-4 predictions to measurements from the B&W critical experiments, and

2) comparison of SIMULATE-3 to CASMO-4 calculations for a set of theoretical 2x2 assembly configurations at a number of different burnups. The gadolinia fuel comparison of CASMO-4 to the B&W critical experiments relies on a small number of data points. Despite this limitation, the comparison, along with the SIMULATE-3 to CASMO-4 comparisons, demonstrates the licensees ability to satisfactorily reconstruct pin power distributions with the CASMO-4 and SIMULATE-3 codes.

The licensee uses the combined assembly average power uncertainties and pin power uncertainties to calculate FH, Fq, and Fz uncertainty factors for LEU fuel and gadolinia fuel. The peaking factor statistically combined uncertainties are used for analysis of reload designs for reactor cores containing gadolinia fuel. The calculation of these uncertainty factors are found to be acceptable for both LEU and gadolinia fuel.

Based on the evaluation of TR DPC-NE-1005-P, Revision 1, as delineated above, the CASMO-4/SIMULATE-3 methodology is deemed acceptable for calculating steady-state physics parameters for use in reload design analyses for McGuire and Catawba reactor cores containing gadolinia fuel.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (73 FR 5218). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

Based on the technical evaluation of Revision 1 to DPC-NE-1005-P, as stated above, the NRC staff finds the proposed changes to the Catawba and McGuire FOLs and UFSARs acceptable.

The NRC staffs approval is subject to, and limited to, the range of fuel configurations and core design parameters as stated and referenced in Dukes November 12, 2007, submittal.

Introduction of significantly different or new fuel designs will require further validation of the above stated physics methods for applications to Catawba and McGuire by the licensee, and will require approval by the NRC staff.

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

The proposed changes to the FOLs and UFSARs based on Revision 1 to DPC-NE-1005-P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, have been reviewed by the NRC staff. Revision 1 of DPC-NE-1005-P presents results of bench-marking studies comparing CASMO-4/SIMULATE-3 reactivity and power distribution predictions to measurements from operating reactors and critical experiments. The report details the methodology used to calculate uncertainties for reload core designs with LEU fuel containing gadolinia.

7.0 REFERENCES

1. T. C. Greer, Duke Energy Carolinas, LLC, letter to the U.S. NRC, November 12, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML073180331).

2. T. C. Greer, Duke Energy Carolinas, LLC, letter to the U.S. NRC, April 8, 2008 (ADAMS Accession No. ML081020280).

3. T. C. Greer, Duke Energy Carolinas, LLC, letter to the U.S. NRC, May 4, 2007(ADAMS Accession No. ML071370090).

4. J. R. Morris, Duke Energy Carolinas, LLC, letter to the U.S. NRC, March 31, 2005 (ADAMS Accession No. ML051010313).

5. R. E. Martin, U.S. NRC, letter to Duke Energy Carolinas, LLC, August 20, 2004 (ADAMS Accession No. ML042370178).

6. NIST/SEMATECH e-Handbook of Statistical Methods, http://www.itl.nist.gov/div898/handbook/, October 2008.

7. Sandia Corporation, Factors for One-Sided Tolerance Limits and for Variable Sampling Plans, SCR-607, pp. 46-54 (Table 2.4), March 1963.

8. ANSI, Assessment of Assumption of Normality (Employing Individual Observed Values),

ANSI-N15.15-1974, Washington, DC.

9. U.S. NRC, An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification, Regulatory Guide 1.126, Revision 1.

10. Babcock and Wilcox Fuel Company, Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, April 1984.

Principal Contributors: A. Attard, J. Woods Date: November 12, 2008